XYLENE POWER LTD.

FAST NEUTRON REACTOR (FNR) DESIGN

By Charles Rhodes, P.Eng., Ph.D.

INTRODUCTION:
SMALL MODULAR LIQUID SODIUM COOLED FAST NEUTRON REACTORS (SM-FNRs) ARE REQUIRED FOR A SUSTAINED SUPPLY OF NUCLEAR POWER, FOR DISPOSAL OF HIGH LEVEL NUCLEAR WASTE, FOR COMPLETE HARVESTING OF ENERGY FROM URANIUM AND THORIUM, FOR LOAD FOLLOWING ELECTRICITY GENERATION AND FOR DISTRICT HEATING.

Large operating and maintenance cost savings are realized by eliminating any necessity for continuous on-site technical staffing of FNRs. If an equipment fault or other problem occurs the SM-FNR defaults to either a warm or cold shutdown state as appropriate. The cold shutdown systems incorporate large safety margins to ensure proper operation in the most adverse circumstances.

Fabrication economy is achieved via use of hundreds of identical prefabricated truck transportable modules which fit together on the reactor site like a child's interlocking building blocks. Each module is factory assembled, tested and warehoused. There is a high degree of certainty with respect to the delivery schedule, commissioning schedule and the installed cost. Due to use of uniform modules maintenance costs and spare parts issues are minimized. Due to extensive equipment redundancy most maintenance problems and safety system performance checks do not require full reactor shutdown.

PURPOSE:
This web page sets out preliminary design calculations for a 1000 MWt Small Modular (SM) Fast Neutron Reactor (FNR) which is assembled from factory fabricated truck transportable modules. The purpose of these calculations is to provide a starting point for the detailed design of a Small Modular Fast Neutron Reactor (SM-FNR) technology that can be mass produced and widely deployed. The proposed fuel cycle enables maximum possible energy harvesting from available spent water cooled reactor fuel, natural uranium and in the future, natural thorium.

CONCEPT:
A 1000 MWt breeding FNR is an assembly of 892 vertical axis fuel bundles immersed in a 21 m diameter pool of liquid sodium. The reactor has an inherent negative slope thermal power versus temperature characteristic which normally keeps the liquid sodium pool surface at a nearly constant temperature in the range 440 deg C to 450 deg C. The reactor power is controlled by controlling the rate of heat extraction from the liquid sodium pool.

Each fuel bundle is nominally 6 m tall X 0.4 m wide X 0.4 m long and is supported by 4 X 3 m legs. During installation each active fuel bundle is fitted with an additional 3 m tall chimney which enhances its natural circulation of liquid sodium. The reactor core has a pancake shape 11.2 m diameter X 0.4 m thick. The reactor core is surrounded by a neutron absorbing blanket about 1.6 m thick which in turn is surrounded by a liquid sodium guard band 2.5 m to 2.9 m thick.

Each fuel bundle is assembled off-site and is transported to the reactor site in a lead shielded transport container carried by a flat bed truck. At the reactor site a crane picks up the transport container, rotates it from the nearly horizontal to vertical position and lifts it up to mate with the reactor's air lock port. After mating to this port the fuel bundle is lowered into the liquid sodium pool. Then the airlock port is closed and the fuel bundle transport container is returned to the transport truck.

Below the air lock port is a low level gantry crane running on I beams.

The reactor enclosure interior ceiling is about 4 m above the liquid sodium surface. This ceiling height is sufficient to provide clearance for the intermediate heat exchange piping and the low level gantry crane and minimizes the potential sloshing of liquid sodium in a severe earthquake.

The low level gantry crane is used to move the fuel bundle to its intended position in the reactor. Moving a fuel bundle entails lifing it vertically about 2 m, moving it horizonally while it remains immersed in the primary liquid sodium, rotating it about its vertical axis as necessary and lowering it into position to mate with a pool floor mounted vertical support tube. The low level gantry crane has a laser positioning system which enables it to achieve a fuel bundle position accuracy of about +/- 1 mm. The fuel bundle support tube projecting above the pool bottom has a tapered slide fittings so that it will reliably mate with the fuel bundle.

After an active fuel bundle is in position the low level gantry crane is used to add the fuel bundle's chimney and indicator tube and fuel bundle assembly reinforcements.

After a fuel bundle has been in service for about 30 years the low level gantry crane removes the fuel bundle's chimney and indicator tube and then moves the fuel bundle to a cooling position at the outer rim of the reactor out of the main neutron flux. There it remains for about 6 years to allow cooling by fission product natural decay. Then the fuel bundle is extracted from the reactor enclosure by reversing the fuel bundle insertion process. A road truck transports the fuel bundle to the fuel reprocessing site.

This operating arrangement simplifies nuclear power station design, reduces the nuclear generating station area requirement and reduces costs.

The reactor enclosure roof is supported by steel lattice girders. This reactor enclosure design minimizes the required argon cover gas volume which in turn minimizes the sizes of the bladder tanks required to accommodate thermal expansion/contraction of the argon cover gas with changes in liquid sodium temperature. The bladder tanks are in silos and are thermally protected by heat exchangers which limit the argon temperature flowing into the bladder tanks.

All of the low level gantry crane operations are conducted after the reactor fission power has been reduced to zero and the liquid sodium has cooled to a temperature of about 120 degrees C. This temperature and the reduced radiation level are tolerable by the crane control electronics. Scheduled fuel bundle exchanges are infrequent (once per six years) which minimizes reactor downtime.

After the fuel bundles are in their desired positions and the active fuel bundle indicator tubes are connected, the gantry crane is moved out of the way of the overhead monitoring instrumentation, the gantry crane electronic package is moved to a thermally and radiation protected environment and reactor operation is resumed. To initiate normal reactor operation liquid sodium powered hydraulic pistons raise or lower the control portions of each active fuel bundle to position setpoints which give the desired liquid sodium discharge temperature for each fuel bundle.

While the reactor is operating ceiling mounted scanners monitor for each fuel bundle the insertion of the fuel bundle control portion, the liquid sodium discharge temperature and the fission related gamma ray flux. This data is used to optimize the fuel bundle control portion position setting.

The use of a low level gantry crane for fuel bundle and intermediate heat exchange bundle positioning substantially reduces the required reactor enclosure size, reduces argon cover gas volume requirements and enables positioning of the fuel bundle monitoring instrumentation relatively close to the tops of the indicator tubes. The use of multiple independent intermediate heat transport systems allows the reactor to continue safe operation in the presence of an intermediate heat exchange bundle failure.

There are redundant equipment insertion/extraction air locks on opposite sides of the reactor.

During periods when a fuel bundle or an intermediate heat exchange bundle is being replaced the primary sodium is cooled to 120 degrees C. The gantry crane moves the bundle to be removed to a position directly below an overhead insertion/extraction airlock. An external hoist then lifts the bundle into the transportation container mated to the air lock.

DESIGN OBJECTIVE:
The object of this web page is to develop a design for a modular FNR with a thermal power rating of about 1000 MWt (320 MWe). This reactor is not dependent on close proximity to a large water body such as a river, lake or ocean for either heat sinking or component transportation.

This design provides a reactor with the following general features:
1) Nothing is neutron activated except the fuel bundles and the primary sodium. The neutron activated primary sodium-24 has a half life of about 15 hours.
2) The primary sodium naturally circulates.
3) The safety systems are all passive. The reactor is "walk-away safe." On loss of thermal load with no fuel geometry change the primary sodium spontaneously seeks its safe stable temperature.
4) On loss of control power the reactor fails to cold shutdown. There is sufficient natural circulation of both primary and intermediate sodium to reliably remove fission product decay heat.
5) With appropriate fuel recycling the production of long lived nuclear waste is at least 1000 fold less per kWhe than for a CANDU reactor.
6) With fuel recycling the ongoing natural uranium requirement is about 100 fold less per kWhe than for a CANDU reactor.
7) The reactor is designed for safe unattended operation with abnormal conditions reporting to a remote monitoring station.
8) Either local safety alarms or a remote command can trigger a reactor cold shutdown and lockout until skilled maintenance personnel can assess and correct the problem.
9) Each reactor has multiple independent heat transport and electricity generation systems. On failure of any one of these systems the others can automatically increase their power outputs to meet the load.

MODULE SIZE:
Each module is of a size and weight that lends itself to inexpensive road truck transport. The practical implication of this concept is that the length of any single long rigid component must be less than 20 m (60 feet) and the outside diameter of any cylindrical object must be less than 5 m (16.4 feet) and the total module weight, including any required shielding, must be less than 100 tons. Depending on the location of a reactor site there may be even tighter dimensional or weight constraints imposed by highway overpasses, bridges, railway tunnels, or air transport. Transportation costs decrease substantially if the maximum component length is less than 52 feet (15.8 m), the maximum cylindrical component diameter is less than 14 feet (4.5 m) and the maximum component weight is less than 70 tons. Thus, if possible, the smaller module dimension and weight constraints should be adopted.

MODULE MANUFACTURE:
The modules are manufactured in a factory, tested, warehoused and shipped to the reactor site ready to install. This process increases quality and provides public utility companies cost and delivery schedule certainty.

DESIGN CONCEPT
The design concept is to assemble a power reactor using a large number of standard size truck portable fuel bundles and intermedite heat exchange bundles to achieve 1000 MWt. Each standard fuel bundle is 0.4 m long X 0.4 m wide X 6 m high. It has 3 m long bottom legs that make the overall fuel bundle height about 9 m. A 3 m high fuel bundle chimney is added on site to achieve an installed overall fuel bundle height of 12 m. The assembly of fuel bundles is completely surrounded by a guard band of liquid sodium almost 3 m thick which absorbs any neutrons that excape from the fuel bundle assembly.

The fuel rods are of two types, blanket fuel rods that absorb neutrons and core fuel rods that both emit and absorb neutrons. The blanket fuel rods are composed of 90% U-238 and 10% Zr. The function of the blanket rods is to capture surplus neutrons emitted by the core rods and to produce more Pu-239/Pu-240. During fuel reprocessing the Pu-239/Pu-240 that is extracted from the blanket rods is used to make new core fuel rods.

The fuel bundles are of two types. There are active fuel bundles which contain both core fuel rods and blanket fuel rods and there are passive fuel bundles which contain only blanket fuel rods.

The active fuel bundles each have two portions, a vertically sliding control portion and a fixed surround portion. The fuel tubes comprising the active fuel bundles have four sections: a fuel tube plenum, an upper blanket, a core region and a lower blanket. The active fuel tube plenum is 2.4 m high. The two blankets are each 1.6 m thick and contain fuel rods initially consisting of 90% U-238 and 10% Zr. The core region is 0.35 m to 0.40 m thick and contains fuel rods initially consisting of 70% U-238, 20% Pu and 10% Zr. The reactivity of an active fuel bundle is set by moving the active fuel bundle control portion vertically with respect to its fixed surround portion using a liquid sodium hydraulic piston actuator located within the fuel bundle support tube.

For safety each active fuel bundle control portion actuator has an independent control system which takes into account the vertical position of the active fuel bundle control portion, the corresponding active fuel bundle gamma flux and the corresponding active fuel bundle discharge temperature. There are two entirely independent cold shutdown systems either of which can shut down the reactor if any active fuel bundle control system is not functioning properly. Any problem fuel bundles can be individually isolated and kept off so as to allow continued operation of the balance of the reactor.

In the event of loss of control power the hydraulic actuators lose hydraulic pressure and gravity causes all the active fuel bundles shut down.

At low thermal power the liquid sodium intermediate coolant also flows by natural circulation. At higher thermal power the liquid sodium intermediate coolant is circulated by electromagnetic induction pumps. Zero thermal power is achieved by reversing the electromagnetic induction pumps sufficiently to prevent natural circulation of the intermediate liquid sodium.

In normal reactor operation the lowest permitted primary liquid sodium temperature is 340 degrees C to prevent precipitation of any entrained NaOH on filters and heat exchange surfaces. The highest permitted primary liquid sodium temperature is 450 degrees C to avoid fuel tube material phase changes that commence at about 475 degrees C. By operating in the specified temperature region with appropriate fuel tube material fuel tube swelling is delayed and the fuel working life before reprocessing is set by Pu concentration decay and fission product accumulation. A fuel burnup of at least 15% per fuel cycle is projected.

In normal reactor operation, once the active fuel bundle control portions are properly positioned, apart from the primary sodium filter system and the hydraulic pressure system, there are no mechanical moving parts in either the primary or intermediate liquid sodium circuits. The primary sodium naturally circulates and the primary liquid sodium temperature is regulated by the change in reactor reactivity with temperature.

The reactor thermal power is controlled by controlling the intermediate liquid sodium circulation rate via an electric induction pump. The intermediate liquid sodium return temperature is indirectly controlled by the pressure regulating valve in the steam generator which maintains a steam generator internal pressure of 11.25 MPa corresponding to a steam generator water temperature of 320 degrees C. By modulating the intermediate liquid sodium circulation rate with a solid state inverter the reactor thermal power output to the steam generator can follow rapid changes in electricity grid power requirements.

Disadvantages of liquid sodium are:
1) Dangerously incompatible with water;
2) Flammable in air at its operating temperature;
3) Must be kept above 98 degrees C during maintenance shutdown periods to keep it in its liquid phase.
4) Its high thermal coefficient of expansion can potentially lead to large pipe and tube thermal stresses at temperatures below the melting point of liquid sodium.

SLUDGE ISSUES:
1) Stable sodium is Na-23. If stable sodium absorbs a slow neutron it becomes Na-24. Na-24 naturally decays with a half life of 15 hours to become stable Mg-24. Magnesium has a melting point of 650 degrees C which is above the highest normal FNR liquid sodium operating temperature. The Mg-24 is denser than Na-23 and settles to the bottom of the primary sodium pool where it forms a sludge.
2) Sodium-23 impacted by fast neutrons will form stable F-19 and stable He-4. The F-19 will immediately chemically react with the Na to form NaF which has a melting point of 993 C. The NaF settles in the liquid Na to form a sludge.
3) Any accidental contact between hot liquid sodium and air will lead to formation of Na2O and NaOH which have melting points of 1132 C and 318 C respectively. The Na2O settles to form a sludge.
4) The sludge components should be constantly removed by filtering. that should be constantly removed by filtering to prevent sludge materials from depositing in the FNR cooling channels or on the intermediate heat exchange surfaces.
5) The sludge accumulation is deep enough below the fuel tubes that there are no neutrons incident upon the sludge accumulation. Hence we need not be concerned about formation of Mg-25 or Mg-26.
6) The filter system inlet should be at the lowest point in the primary liquid sodium pool.
7) The NaOH can only be filtered out when the reactor is in cold shutdown.
8) During normal reactor operation the liquid sodium is everywhere kept above 320 C to prevent the NaOH depositing on cool heat exchange surfaces. The liquid NaOH will tend to collect at the lowest point in the primary liquid sodium pool.
9) The lowest normal operating temperature of the intermediate liquid sodium should be 330 degrees C to prevent NaOH precipitating on steam generator heat exchange surfaces and to prevent scouring those surfaces.
10) There must be a mechanism for periodic filtering NaOH out of the secondary loops after discharge from the steam generators.
11) We need to normally operate the system with the secondary liquid sodium discharge from the steam generator at 320 C to prevent any internal NaOH deposition or scouring within the steam generator tubes. That requirement in combination with the saturated vapor pressure of water at 320 degrees C implies an 11.25 MPa working pressure for the steam generator and an 11.5 MPa internal working pressure rating for the intermediate heat exchanger.

These and other properties of sodium such as its viscosity together dictate numerous aspects of liquid sodium cooled FNR power plant design.

STEAM GENERATOR WORKING PRESSURE:
The temperature of the water in the steam generator is kept at 320 deg C = 608 deg F to prevent precipitation of NaOH, which melts at 318 degrees C, in the intermediate cooling loop. The corresponding saturated vapor pressure for water is:1637.3 psia
= 1637. 3 psia X 101 kPa / 14.7 psia X 1 MPa / 1000 kPa
= 11.249 MPa
This pressure is maintained in the steam generator by the steam generator pressure regulating valve, the discharge from which drives the steam turbine.

For safety the steam generator water side must be designed for a yield pressure of at least 36 MPa, must be pressure tested at 18 MPa and must be fitted with a pressure relief valve that trips at 12.0 MPa.

INTERMEDIATE SODIUM CIRCUIT PRESSURE RATING:
For safety, in the event of a steam generator tube rupture the direction of fluid flow through the rupture must be from the secondary sodium circuit into the water-steam, not vice-versa. Thus the liquid sodium intermediate circuits must be rated for a working pressure of 12 MPa. Hence for safety the intermediate sodium circuits must be pressure tested at 18 MPa. For safety the intermediate sodium pressure rated components should all have a Specified Minimum Yield Stress (SMYS) pressure rating of 36 MPa.

MAXIMUM PRIMARY SODIUM TEMPERATURE:
The maximum primary liquid sodium temperature at full load is chosen to be 440 degrees C so that, allowing for a 15 degree C temperature drop across a fuel tube wall the normal maximum fuel tube material temperature is 455 degree C. This temperature choice is made to prevent a Fe-Cr fuel tube material phase transition which can occur at temperatures in excess of 460 degrees C. Note that at no load the primary liquid sodium temperature can reach 450 to 455 degrees C but at no load the temperature drop across the fuel tube wall is zero. It is important to precisely position the active fuel bundle control portions so that all the interior fuel bundle discharge temperatures are equal. The control portions of the outer ring of active fuel bundles must be positioned to meet the safe shutdown criteria.

At no load the intermediate sodium minimum temperature will theoretically drop down to about 320 degrees C. It may be necessary to operate each turbogenerator at a minimum load to maintain generator to grid synchronization.

SODIUM RELATED FNR DESIGN ISSUES:
1) Except in the steam generator equipment spaces, water and liquid sodium should not be present in the same building because when water and liquid sodium contact hydrogen is rapidly released along with sufficient heat to trigger spontaneous hydrogen ignition. Hydrogen is flamable in air over a wide range of hydrogen-air ratios.

2) The fluid used for transporting heat away from the primary liquid sodium pool is non-radioactive liquid sodium pressurized by the inert gas argon. Hence an intermediate heat exchanger tube rupture has little serious consequence other than adding a small volume of non-radioactive sodium to the radioactive sodium pool. In the event of a steam generator tube rupture the object is to immediately vent the steam and to keep the intermediate liquid sodium pressure slightly above the water pressure (steam pressure) to minimize liquid sodium leakage through the rupture and to prevent water or steam entering the intermediate liquid sodium circuit.

3) If the intermediate sodium circuit is not designed for high pressure operation then on a steam generator tube rupture the pressure in the secondary sodium circuit will instantly rise and will likely rupture the corresponding intermediate heat exchanger. This problem may be amplified by liquid sodium hammer in the intermediate sodium circuit. This type of failure could have very serious consequences. The simple solution is to design the intermediate sodium circuit to operate at a high pressure.

4) When sodium-24 decays to become Mg-24 it emits 1.389 MeV electrons and emits 1.369 MeV gamma rays. Hence manual service work in the proximity of the radioactive primary liquid sodium must be delayed for about a week (11 Na-24 half lives) after reactor shut down to allow the Na-24 to naturally decay. To minimize maintenance downtime there are few moving parts within the reactor enclosure and normal service work in the reactor enclosure is done via robotic equipment. The likely service issues within the reactor enclosure are fuel bundle repositioning, intermediate heat exchange bundle replacement, fuel bundle control portion actuator system service and primary sodium filter system service.

5) A major issue with use of liquid sodium as a coolant and heat transport medium is that the density of liquid sodium is 0.927 X (density of water) and the (heat capacity / kg of liquid sodium) is about (0.34 X the heat capacity / kg of water).

6) In order for liquid sodium to convey heat with approximately the same size pipes and the same fluid velocity as would be used for water the heat transport loop temperature differential must be increased about four fold. That increased loop temperature differential causes significant thermal stress relief design issues in the intermediate heat exchanger and in the steam generator. The heat exchange systems and the intermediate sodium flow rate must be designed to minimize the temperature drop across the intermediate heat exchange tube walls and across the steam generator tube walls.

7) Liquid sodium is used as the intermediate coolant in a fast neutron reactor because sooner or later due to an intermediate heat exchanger tube or manifold failure high pressure secondary liquid sodium will leak into the low pressure primary liquid sodium. There should be a sufficient number of independent isolated secondary liquid sodium circuits to ensure that a single secondary heat transfer circuit fault will not force a total reactor shutdown.

8) Note that the secondary sodium circuits must be designed to accommodate high thermal stress both at the design operating temperature and at temperatures below the melting point of sodium.

9) There must be a system for raising the liquid sodium pool above its melting point of 98 C to enable reactor startup.

10) Every component of the liquid sodium SM-FNR is easily replaceable except the primary liquid sodium pool liners, the brick work and the primary liquid sodium. The inner pool liner should be carefully designed to last for centuries because the cost of a reactor shutdown and transfer of the primary liquid sodium into holding tanks to enable inner liner repair work is very high. In this respect the most critical elements are the welds used to assemble and seal the stainless steel pool liner. The stainless steel pool liner material must be chosen for continuous containment of liquid sodium at 340 C to 455 C.

11) An important issue in FNR design, fabrication and operation is keeping the liquid sodium clean so that over time grit and debris do not obstruct the natural circulation of liquid sodium either through filters or in the narrow cooling channels between the FNR fuel tubes. For protection certainty each active fuel tube bundle is fitted with its own primary liquid sodium filters.

13) Each turbine hall will contain 8 X 10 MWe turbogenerators, steam condensers and related injection pumps. The turbine halls are air ventilated to enable on-going service work. Below the turbine level are the liquid sodium drain down tanks which provide storage volume for the intermediate liquid sodium when an intermediate liquid sodium circiut is being serviced.

14) The equipment containing intermediate sodium should exist in an oxygen depleted atmosphere to prevent a fire if there is a leak of high pressure intermediate sodium. The steam generator equipment spaces must be fitted with sodium carbonate (Na2CO3) equipment for extinguishing a sodium fire. Note that anhydrous Na2CO3 decomposes into Na2O + CO2 at 851 degrees C.

15) The 1 m thick concrete wall between the reactor space and the steam generator equipment and turbine hall spaces is structurally sufficiently thick to provide radiation shielding to allow safe work in the steam generator space while the reactor is operating. This wall has the secondary function of keeping water out of the reactor space and safely absorbing a possible hydrogen explosion in the steam generator space. This wall must be gas tight to prevent CO2 or air entering the reactor space. The gas tight seal is realized via a sheet stainless steel wall covering and bellows sealed pipe feedthroughs.

16) Each steam generator space must be equiped for safely venting hydrogen to the atmosphere through blowout panels in its walls near the ceiling.

17) Each steam generator space needs a chemical system for safe controlled oxygen absorption. eg Na converts to Na2O at a low temperature to limit the reaction rate.

18) The floor between each steam generator space and its companion turbine hall must be gas tight to prevent air or water vapor from the turbine hall entering the steam generator space.

19) Each steam generator equipment space requires an air-tight door to the outside to allow equipment replacement and requires a small airlock for personnel entry-exit. When air is admitted for equipment inspection or replacement all the equipment in that steam generator equipment space should be shut down and cool.

20) The intermediate sodium pipes and the steam pipes will thermally expand and contract. Hence these pipes will move a lot with respect to rigid walls and floors. Hence there will have to be bellows type fittings around the pipes to gas seal to the walls at the points where the sodium pipes pass through the walls between the reactor space and the steam generator spaces and at the points where the steam pipes pass through the floor of the steam generator space to reach the corresponding turbine hall. The steam generators are generally rigidly mounted implying that the intermediate heat exchange bundles in the reactor must be able to move to relieve thermal expansion-contraction stress. Hence adequate intermediate heat exchange bundle mounting clearances must be provided.

21) Due to lack of ventilation the atmosphere in the steam generator spaces will be very hot. Practical work in those spaces will require shutting down (1 / 4) of the reactor capacity so that the pipes and steam generators in the work zone can be cooled to about 120 degrees C. That cooling will also entail removal of presurization from the involved secondary sodium loops. Removal of that pressurization will make work in that steam generator equipment space much safer.

22) The induction type secondary sodium pumps will need circulated oil cooling. Their cooling circuits must be designed so that natural circulation is sufficient to protect these pumps under all conditions. The chosen oil should have a large temperature coefficient of expansion. The cooling towers must include oil coolers. All of the electrical materials located in the steam generator equipment space must be rated for high temperature operation.

23) The reactor space will use two bladders in silos to achieve automatic argon pressure control.

24) The four steam generator spaces will all need automatic oxygen depleted air pressure control systems.

25) The steam generator equipment spaces will all need provision for air cooling prior to service access.

26) Each steam generator equipment space should have a thermally insulated outer perimeter corridor with insulated windows that permit routine visual equipment inspection without requiring personnel access to the equipment space. It should be possible to operate the Na2CO3 fire suppression equipment from this corridor.

FNR SODIUM POOL:
The primary sodium pool for a 1000 MWt FNR is described on the web page titled FNR SODIUM POOL

1000 MWt REACTOR ASSEMBLY:
The assembly of fuel bundles is located in the lower centre of the primary liquid sodium pool between 3.5 m and 9.5 m above the pool inside bottom.

The maximum diameter of the reactor fuel bundle collection is 15.2 m from one straight face to the opposite straight face. The minimum distance from a fuel bundle assembly face to the nearest primary liquid sodium containment wall is about 2.9 m.

There are two more equivalent heat exchange bundle positions that are used for loading and unloading the air locks. Note that in the ring of heat exchange bundles there are two gaps sufficiently wide to permit insertion, removal and replacement of intermediate heat exchange bundles and fuel bundles.

The following diagram shows a side elevation of a 1000 MWt FNR.

The bottom and sides of the primary sodium pool are lined with stainless steel sheet and are thermally isolated from the environment by a 3 m thick layer of porous lava rock.

Not shown are the 1068 X (12.00 inch X 12.00 inch X 2.0 m long) vertical square steel pipes that position, support and stabilize the fuel bundles. The 640 active fuel bundle control portion actuators are located inside some of these square steel pipes. The fuel bundle assembly weight is distributed over the primary sodium pool floor using horizontal rectangular structural steel tubes each 14 inches X 6 inches X 0.625 inches. The fuel bundles are stabilized by being clipped to their nearest neighbours at their top corners.

There may also be a fuel bundle assembly waiste stabilization frame with reinforcing diagonals.

Note the 640 X (7.5 m long indicator tubes) that reach above the top surface of the primary liquid sodium.

Not shown above the primary sodium pool is the electro-optical system that gathers active fuel bundle status data from the indicator tubes.

Note that the intermediate heat exchange tube bundles are baffled and are positioned above the fuel tube bundles to enhance natural circulation of the primary liquid sodium. The primary sodium, after being cooled by an intermediate heat exchange bundle, is ducted down to the bottom of the primary sodium pool.

The maximum diameter of the reactor fuel bundle assembly is 15(0.4 m) = 6.0 m from one octagonal face to the opposite octagonal face. The minimum distance from a fuel bundle assembly octagonal face to the nearest primary liquid sodium containment wall or the nearest intermediate heat exchange component is 2.4 m.

The outer ring of blanket fuel bundles has 44 positions suitable for storing used active fuel bundles out of the main neutron flux, without intruding into the intermediate heat exchange zone, in order to allow for fission product decay while still immersed in sodium. This mounting space is also used for repositioning fuel bundles. Note that the fuel bundles and intermediate heat exchange bundles are transported horizontally and must be rotated into a vertical position while being suspended above the primary sodium pool.

The bottom and sides of the primary sodium pool are lined with stainless steel sheet and are thermally isolated from the environment by a 3 m thick layer of porous lava rock.

There are the 143 X (12.00 inch X 12.00 inch X 2.0 m long) vertical square steel pipes that position, support and stabilize the fuel bundles. The 37 active fuel bundle control portion actuators are located inside some of these square steel pipes. The fuel bundle assembly weight is distributed over the primary sodium pool floor using horizontal rectangular structural steel tubes each 12 inches X 12 inches X 0.5 inches. The fuel bundles are stabilized at their top corners by being clipped to their nearest neighbours.

There are 37 X (7.5 m long indicator tubes) that reach above the top surface of the primary liquid sodium.

FNR TRADEOFFS:
It is contemplated that the reactor core fuel rods are initially 0.35 m long and are contained in 0.500 inch OD X 0.065 inch wall steel tubes positioned laterally on 0.625 inch square centers. In the vertical channels between the fuel tubes liquid sodium coolant flows upwards to remove heat. Implications of this design are a small liquid sodium circulation power (natural circulation), negligible fuel tube erosion, reasonable fuel temperature, reasonable reactor dimensions, an acceptable liquid sodium temperature rise and a reasonable requirement with respect to filtering particulates out of the liquid sodium.

The size of the gap between the steel tubes is a compromise between the requirements for heat transport and the requirements for average fuel density. As the gap between the fuel tubes becomes smaller the problems of primary sodium circulation and of filtering particulates out of the liquid sodium rapidly become larger. As the gap becomes larger the average concentration of core rod fissionable material rapidly increases.

Practical operating experience with the EBR-2 showed that during normal operation formation of fission products causes the core rod cross sectional area to swell by about 33%. Hence the initial reactor core fuel rod diameter is restricted to:
[(steel fuel tube ID) X 0.86].
The only practical ways to increase the average fuel density in the reactor core are to reduce the gap between the steel fuel tubes or to increase the concentration of Pu-239 in the core fuel rods. Note that the initial blanket fuel rod outside diameter can be slightly larger than the initial core fuel rod outside diameter because the blanket fuel rods are less subject to fission product induced swelling.

An important issue in FNR design is neutron conservation. Almost all the excess neutrons emitted by the reactor core zone should be captured by the surrounding 1.6 m thick breeding blanket.

The amount of plutonium readily available from spent CANDU fuel is about:
0.0038 X 54,000 tonnes = 205.2 tonnes. Hence at this time in 2017 in Canada there is only enough plutonium available to start about:
205.2 tonnes / (17.5 tonnes / reactor) = 11.7 FNRs.
It is clear that in FNR planning a very important objective is breeding additional plutonium for starting future FNRs.

INTERMEDIATE HEAT EXCHANGER:
The fuel bundle and intermediate heat exchange bundle installation - removal pathway is along a common corridor.
There are 32 intermediate heat exchange bundles immersed in the liquid sodium which are used to extract heat from the primary liquid sodium pool. Each intermediate heat exchange bundle must be rated for a full load heat transfer of at least:
(1000 MWt) / 32 = 31.25 MWt.

The intermediate heat exchange tube bundles are located between 6.5 m and 12.5 m above the pool bottom. The primary liquid sodium circulates by natural convection. In the middle of the primary sodium pool the top surface of the liquid sodium is about 15.5 m above the pool bottom. During reactor operation the elevation of the top surface of the liquid sodium slightly increases in the center of the pool and slightly decreases at the edges of the pool. The top surface of the liquid sodium is normally about 1.0 m below the pool deck.

There are 32 independent intermediate heat transfer circuits, each consisting of one baffled counter flow intermediate heat exchange bundle, extended 16 inch OD pipes and fittings, a drain down tank, an induction type intermediate liquid sodium circulation pump, a steam generator, a cushion tank and an argon pressure regulation system. Liquid sodium is transferred from the drain down tank into the secondary loop by application of argon pressure in the drain down tank. The intermediate heat exchange tube bundle and the steam generator tube bundle both operate at a working pressure of 11.25 MPa. The intermediate heat exchange tube material is always under tensile stress. The intermediate heat exchanger design and the steam generator design is constrained by the high temperature yield stress and creep properties of the heat exchange tube material and by the tube wall thickness required to withstand high thermal stress at temperatures below the melting point of liquid sodium.

Each intermediate heat exchange bundle is dedicated about 1.5 m of pool perimeter and functions to remove heat from the upper layer of hot liquid sodium. Natural circulation of the primary liquid sodium conveys heat from the nuclear fuel bundle discharge near the center of the pool to the heat exchange bundles near the edges of the pool. The liquid sodium recirculates along the bottom of the pool.

The intermediate heat exchanger secondary fluid is non-radioactive sodium. The heat exchanger secondary fluid conveys the heat to steam generators which are located in adjacent buildings. The secondary sodium is circulated via electromagnetic induction pumps. In further adjacent buildings steam from the steam generators is expanded through turbines. Condensers inside natural draft cooling towers condense the steam. The condensate is pumped at a high pressure (~ 11.25 MPa) back into the steam generators via recuperator heat recovery coils located in the condensers. The condensate injection rate into the steam generators is adjusted to maintain the desired water levels in the steam generators.

At the fuel bundle pool insertion-removal point the external hoist lifting point must be at least 14 m above the liquid sodium surface. The fuel bundles are moved in the liquid sodium with a 10 ton rated gantry crane.

REACTOR BUILDING:
The reactor building above grade outside foot print is 29 m in diameter. Around the perimeter of the reactor building are 4 steam generator spaces. Below the steam generator spaces are the turbine halls. Above grade shielded fresh air intakes in the middle of the reactor building side walls increase the reactor building's width by about 6 m. The fresh air is ducted down to the outside bottom of the primary sodium pool and over the higher suspended roof. Also at the argon-vacuum-air locks are docks for delivery and removal of fuel bundles and intermediate heat exchange bundles. On the top of the reactor building side walls are multiple exhaust fans.

Off-site storage is used for storing sodium drums and for possible fuel bundle assembly/reprocessing.

All of the FNR components are designed for factory fabrication and are sized for easy truck transport with no special provisions for load over width, over height or over weight. The sheet stainless steel panels forming the inner pool liner, outer pool wall, inner metal ceiling, inner above grade metal wall, outer metal ceiling and outer above grade wall must be field welded together. These welds must be air tight and free from defects.

The 2.2 m long heat exchange bundle sections at the edge of the pool are 6.0 m deep, and are supported by front and back vertical pipes. The intermediate sodium pipes are supported by a hanger arrangement that allows thermal expansion-contraction.

In plan view the primary liquid sodium pool has a 2.6 m to 2.9 m wide neutron absorption guard band around the fuel bundle assembly. The spent fuel bundles are stored out of the neutron flux at the edge of the fuel bundle assembly to allow fission product decay. The central 14.4 m diameter area is occupied by reactor active and passive fuel bundles. At the edge of the primary sodium pool and above the fuel tubes the 2.3 m wide perimeter region that is occupied by the intermediate heat exchange bundles. The 2.6 m to 2.9 m liquid sodium guard band between the cooling fuel bundles and the nearest wall or heat exchange bundle assembly protects the walls and heat exchange bundles from cumulative neutron damage. Fuel bundles and intermediate heat exchange bundles enter or leave the reactor building via argon-air locks.

GUARD BAND:
The guard band is a region 2.6 m to 2.9 m wide outside the fuel bundle assembly perimeter. The guard band contains no equipment except the fuel bundle support tubes. The first 1.2 m of reactor blanket prevents most of the neutrons originating in the reactor core from being absorbed by the cooling bundles. There is another 0.4 m of cooling bundle thickness that further absorbs neutrons. There is a further 2.6 m to 2.9 m of sodium to the intermediate heat exchange bundles and the pool wall. The purpose of the guard band is to allow active fuel bundle cooling, extend equipment life and minimize formation of decommissioning waste. Above the tops of the fuel tubes are 6 m of liquid sodium that prevent neutron emission upwards, even when a fuel bundle is raised 2.5 m for movement in the reactor. Below the bottoms of the fuel tubes are 3 m of liquid sodium that prevent neutron absorption by the primary sodium pool floor or sludge on the pool floor.

SPENT FUEL BUNDLE STORAGE SPACE:
Around the perimeter of the reactor is space for storing up to:
176 spent active fuel bundles
out of the main neutron flux. This fuel bundle cooling space is used to allow fission products to decay to a level compatible with fuel reprocessing before a fuel bundle is removed from the primary liquid sodium pool.

FLOATS:
In normal operation the entire top surface of the primary liquid sodium pool is covered by an array of 0.4 m X 0.4 m square steel floats. The purpose of these floats is to minimize the liquid sodium surface area that is exposed to the cover gas atmosphere and to stabilize the relative positions of the indicator tubes. In theory the atmosphere above the liquid sodium is argon. However, from time to time some air will mix with this cover gas, in which event the floats minimize sodium-air chemical reactions which will tend to pollute the liquid sodium.

An important function of these floats is to minimize the fire risk and to minimize sodium vapor condensation on the interior walls, ceiling, gantry cranes and optical scanning equipment. The floats have holes in their centers to allow passage of the indicator tubes. At the perimeter of the reactor are floats that are shaped to cover the sodium over the intermediate heat exchange bundles.

REACTOR CONCRETE ENCLOSURE:
The space immediately above the primary sodium pool is filled with an inert cover gas (argon) that will not chemically react with the liquid sodium. The immediately overhead roof (the lowest roof) must be high enough to allow fuel bundle and heat exchange bundle remote manipulation via the gantry cranes and must be gas tight. The lowest roof is sheet stainless steel and is suspended from the steel lattice roof structure above it. The suspension rods have thermal breaks. The lowest roof operates at about 450 degrees C. On top of the lowest roof is a 1 m thick layer of high temperature rated ceramic fibre insulation (fiberfrax). On top of this insulation is the outer metal roof which is near ambient temperature. On top and outside of the outer metal roof is a 3 m thick space for circulation of cooling air. This space also allows human access for roof service commencing about one week after the reactor is shut down.

Included in the air space is the 2 m thick steel lattice roof support. The roof has two airlocks to permit replacement of fuel bundles and intermediate heat exchange bundles. The purpose of the external roof is to:
1) Exclude rain water and snow melt water;
2) Provide severe storm protection for the FNR.
3) Provide physical protection for the FNR from either overhead or grade level physical attack;
4) Contain and smother a sodium fire;
5) Provide reserve radio isotope confinement.

In plan view the inside dimensions of the bottom of the concrete walls are: 27 m ID, 29 m OD.

Each fuel bundle is brought in vertically from a truck transport container via a argon-air lock in the reactor enclosure roof. When the fuel bundle is vertical its bottom is immersed in liquid sodium. It is then moved horizontally through the liquid sodium to its desired rest position.

Then the volume of concrete required is given by:
(volume of base) + (volume of straight walls)
= 661 m^3 + 2155 m^3
= 2816 m^3 concrete
which does not include the concrete requirements for the steam generator and turbogenerator spaces or the cooling towers.

The external steel roof must be covered by a durable waterproof membrane that is easily serviceable. This roof must be able to exclude rain water under the most adverse circumstances, including violent storms, tornados, long term corrosion and deliberate aerial attack. Ensuring a long term reliable roof is a major issue in safe liquid sodium cooled FNR implementation. If the roof is formed from concrete it might use interior square recesses, similar to the Roman Pantheon, to reduce its weight without seriously impacting its strength.

There should also be an an argon based fire suppression system sufficient to prevent sodium combustion in the event of a major roof failure or in the event of a heat transfer circuit pipe rupture.

LIQUID SODIUM POOL CONSTRUCTION:
There are no through holes in either the side walls or the bottom of the liquid sodium pool. The cover gas above the pool is inert (argon). The ceiling above the pool center line is about 3 m above the liquid sodium surface. The reactor enclosure ceiling is insulated and the gas above the pool is maintained slightly above the pool surface temperature (~ 445 deg C) to prevent sodium vapor condensation on the ceiling and on the monitoring instrumentation window.
A rotorary gantry can be used to maniplate the intermediate heat exchange bundles.

The pool floor holds a steel frame with a 0.400 m square grid of ~ 12 inch square ~ 2.0 m deep plugs that are used to position, support and stabilize the fuel bundles. A small central hole in each socket provides controlled amounts of high pressure liquid sodium for active fuel bundle control portion vertical positioning.

The primary liquid sodium pool has a double wall and a double sub-floor formed from sheet stainless steel for primary sodium containment certainty. The inner and outer walls are separated by a 3 m thickness of saw cut lava rock that limits heat loss through the pool walls and floor and limits the decrease in primary sodium pool depth in the event of a failure of the inner wall liner. The liquid sodium pool must be sited at sufficient elevation that the liquid sodium will never be exposed to flood water or ground water. The ground surrounding the pool must be sufficiently above the local water table that in the event of a major earthquake or other event that ruptures the pool inner, the pool outer wall and the concrete wall the contained radio active sodium still cannot go anywhere or react with significant quantities of ground water.

If the reactor is located on a hill top the primary liquid sodium pool requires a below grade excavation of at least:
Pi (29 m / 2)^2 X 21 m deep
= 13,871 m^3.

The pool floor must be well supported because it carries the entire weight of the liquid sodium plus the weight of the fuel bundles and their control rod apparatus plus the weight of the immersed heat exchangers and their associated piping plus the weight of the fuel bundles in storage plus the weight of the pool walls and floor, including the 3 m thickness of lava rock insulation.

After pouring the concrete foundation slab the straight concrete walls are erected first. Then the lava rock and pool liner are placed.

STABLE STRATIFIED PRIMARY LIQUID SODIUM:
The reactor is designed to operate with hot liquid sodium (440 deg C) occupying the upper 3.0 m to 10 m of the pool depth, and with cooler liquid sodium (340 deg C) occupying the bottom 5.5 m to 12.5 m of the pool depth. In the:
7 m
between these two extremes the temperature in the primary liquid sodium pool changes depending on reactor loading. The design temperature difference between the top and bottom of the primary liquid sodium pool is 100 degrees C.

The elevation of the center of the transition region between the hot liquid sodium on top and the cooler liquid sodium on the bottom is referred to as the transition elevation. Due to an insulated baffle the transition level is different at the intermediate heat exchanger than at the reactor. If the transition elevation changes natural circulation through the intermediate heat exchanger primary and through the reactor change, which tends to restore the transition elevation to its desiqn level.

The intermediate heat exchanger inputs hot primary liquid sodium from near the liquid sodium top surface and discharges cool primary liquid sodium near the pool bottom.

Note that to operate at a high thermal power requires high liquid sodium natural circulation which requires a high primary liquid sodium pool top to bottom temperature differential. Hence the reactor power rating is dependent on the steam generator design being consistent with this high liquid sodium temperature differential.

Material property limitations limit the primary liquid sodium top surface operating temperature to a maximum of 450 degrees C. However, more generally the reactor power is limited by fuel tube, heat exchanger and steam generator material and performance constraints. In normal operation the temperature at the top of the primary sodium pool is 440 C and the temperature at the bottom of the primary sodium pool is about 340 C.

PRIMARY LIQUID SODIUM FLOW PATH:
The primary liquid sodium flows along a coaxial vertical loop path. Flow starts at the bottom center of the primary liquid sodium pool. The liquid sodium rises between the reactor fuel tubes. As the liquid sodium rises it absorbs heat, expands and becomes less dense and hence more buoyant relative to the surrounding cooler liquid sodium in the blanket and guard band. At the pool top surface the liquid sodium flows radially along the pool top surface toward the perimeter of the pool.

Near the perimeter of the pool the liquid sodium flows down thermally isolated vertical ducts containing the single pass of vertical intermediate heat exchange tubes.

DESIGN FOR FNR SAFETY:
This FNR design uses high pressure liquid sodium intermediate heat transport loops. This heat transport fluid choice places two pressure rated metal barriers between the high pressure water/steam and the low pressure radioactive primary liquid sodium. The pressure in the intermediate heat transport loops is maintained by high pressure argon in the heat transport loop cushion tanks. This argon pressure is automatically controlled to track the steam pressure in the relevant steam generator.

In the event of a steam generator tube failure the high secondary sodium loop pressure prevents water from entering the secondary sodium circuit. In the event of an intermediate heat exchanger tube failure the liquid sodium from only one of the 32 separately isolated secondary circuits leaks into the primary sodium pool with little practical consequence. The argon in the leaking secondary sodium circuit's cushion tank is chemically inert and will not chemically react with either hot water or hot liquid sodium.

The surface of the primary liquid sodium pool is covered by 0.4 m X 0.4 m square steel floats which minimize sodium oxidation or combustion in the event that oxygen leaks into the sodium pool's argon cover gas. In the event of a significant air leak the liquid sodium is cooled as fast as possible to below 200 degrees C which is its threshold for combustion in air.

The primary liquid sodium pool is located sufficiently above the local water table that it will never be exposed to flood water.

The primary liquid sodium naturally circulates. This natural circulation system avoids many practical complications and reliability issues relating to pumped circulation of the primary liquid sodium.

The fuel bundles are configured such that the sliding control portion moves vertically with respect to the fixed surround portion.

RESERVE SODIUM STORAGE:
Sooner or later it will be necessary to do maintenance work on the sodium pool inner wall. To do such work it will be necessary to pump the primary sodium out of the pool. Hence there must be container storage volume with sufficient capacity to hold the entire volume of primary liquid sodium while the maintenance work is being carried out. For maintenance flexibility there must be a practical means for transferring liquid sodium into and out of such containers. eg To store 5000 m^3 of sodium may require about 25,000 steel drums.

THERMAL POWER CONTROL:
The reactor core zone attempts to maintain its own temperature setpoint of 440 degrees C. With the control bundles 1.0 m withdrawn the corresponding setpoints are intended to be less than 0 C to ensure total reactor shutdown.

As the liquid sodium in the reactor core zone warms up it thermally expands increasing neutron diffusion out of the core zone and hence stopping the nuclear chain reaction in the core zone. Then the only heat produced is fission product decay heat. Provided that the active fuel bundle control portion insertion is correct and that there is adequate decay heat removal the liquid sodium discharge temperature from that fuel bundle will always stabilize at its design temperature. If heat is removed from the liquid sodium faster than decay heat is produced the liquid sodium temperature decreases causing the liquid sodium density to increase. This liquid sodium density increase reduces neutron diffusion out of the core zone which restarts the nuclear chain reactions.

Care must be used to ensure that every fuel bundle remains within its safe and stable thermal control range. If the fuel is too rich or if the control bundle insertion is too great the fuel in the core zone can potentially get too hot and melt. The fuel bundle insertion rate is mechanically limited to ensure against prompt neutron criticality. Hence the discharge temperature of each fuel bundle and the corresponding gamma power emission are individually monitored. The active fuel bundle control portion insertion is precisely controlled.

The active fuel bundle control portion positions and the fuel bundle discharge temperatures are indicated by indicator tubes that project above the surface of the liquid sodium. The horizontal position of each indicator tube is stabilized by its chimney and by its 0.400 m X 0.400 m steel float.

THERMAL SHUTDOWN:
If the control bundles are properly positioned, as the reactor's external thermal load decreases the primary liquid sodium temperature rises and the primary liquid sodium thermally expands causing an increase in neutron diffusion out of the reactor core zone and hence a reduction in reactor heat output. When a fuel bundle discharge temperature exceeds its setpoint the reactor core zone becomes subcritical and the fission reactions in that fuel bundle totally cease.

Note that at all times the external thermal load must be sufficient to remove fission product decay heat.

REACTOR THERMAL POWER MODULATION:
The reactor thermal power output is modulated by modulating the secondary sodium flow rate. As the secondary sodium flow rate decreases the thermal power delivered to the thermal load decreases. The steam generator must be designed to accommodate the changing secondary liquid sodium flow. The pressure regulator on the steam generator effectively sets the secondary liquid sodium return temperature by modulating the steam discharge valve to maintain the steam pressure in the steam generator at about 11.25 MPa.

REACTOR TRIP CONDITION:
Since the working pressure of the secondary sodium heat transport system is 11.5 MPa the steam pressure must always be kept under 11.5 MPa. The reactor must be shut down if for any reason the steam pressure exceeds 11.5 MPa.

STRESS RELIEF:
In this equipment arrangement the secondary sodium circuit is designed to safely operate with an internal working pressure of up to 12 MPa. The net working pressure stress on the steam generator tubes is minimized by controlling the secondary sodium pressure to be slightly above the steam pressure. Intermediate heat exchanger material thermal stress is minimized by the use of a counter flow heat exchange configuration which limits the temperature difference across the steam generator tube walls and hence limits the tube material thermal stress. These tubes normally operate with an internal sodium pressure of about 11.5 MPa.

POWER BREEDER REACTOR CONCEPT:
The power breeder reactor contemplated herein has one octagonal shaped core zone, 11.2 m in diameter X (0.35 m to 0.40 m) high that is sandwitched between two breeding blanket zones each 1.6 m high. The perimeter of this stack is surrounded by a 1.6 m thick breeding blanket. The whole is in turn surrounded by a 3 m thick layer of liquid sodium for complete neutron absorption.

The reactor is an assembly of 892 fuel bundles. Each active fuel bundle is an assembly of 476 adjacent vertical fuel tubes. Each fuel tube contains a vertical stack of fuel rods.

One neutron per Pu-239 fission is required to sustain the fission chain reaction. One neutron per Pu-239 fission is required to sustain the plutonium Pu-239 production required to provide future fuel for this reactor. Approximately 0.5 neutrons per Pu-239 fission are lost to various unproductive neutron absorption processes in sodium and steel. The remaining 0.6 neutons per Pu-239 fission are used for breeding additional plutonium for starting other breeder reactors.

The 0.533 m high breeding blanket fuel rods are made from 90% uranium and 10% zirconium.

TUBE AND PIPE WALL THICKNESS:
One of the design issues with liquid sodium is that the containing tubes and pipes must have sufficient wall thickness to safely absorb the stresses that can occur during melting of the sodium. Consider a round steel pipe which is plugged at both ends and which is completely full of solid sodium at room temperature.

Note that fuel tubes are subject to wall deterioration due to fast neutron damage to which the intermediate sodium circuit is not subject.

For reasons related to the technology of steel tube manufacture only a few manufacturers can meet the required (Wp / Dp) ratio for 0.500 inch OD tube.

FUEL AGING ISSUES:
The nuclear fuel material within the reactor core and blanket is in the form of rods. The core rods are metallic and are initially 0.35 m long but in use gradually swell to:
(1.0 / .86) X 0.35 m = 0.40 m long.

The 0.533 m long blanket rod sections are formed from an alloy consisting of 90% U and 10% Zr.

Within the reactor core Pu-239 and other actinide fission reactions produce fission products, some of which have high neutron absorption cross sections. Simultaneously, as the fuel and blanket rods absorb surplus neutrons from the Pu-239 fission reactions the U-238 atoms gradually transmute into more Pu-239 and a spectrum of trans-uranium actinides.

Periodically, at a time interval known as the fuel cycle time, the core and blanket rods are removed and reprocessed. The effect of this reprocessing is to extract fission products, to move plutonium and transuranium actinides generated in the blanket rods into new core rods and to replace the lost blanket rod mass with an equal mass of depleted U, which may be also be obtained from spent CANDU fuel. A typical fuel cycle time is about 30 years. About 3.3% of the fuel bundles can be reprocessed every year so the average reactor performance does not significantly change with time and the fuel reprocessing is nearly continuous. The scheduled annual reactor shutdowns are only for a few hours to permit repositioning and exchange of fuel bundles.

FAST NEUTRON POWER REACTOR DESIGN CONCEPTS:
Most of the power FNR design concepts have been extensively tested in small research reactors such as the EBR-2. However, the EBR-2 had less than 10% of the THERMAL power rating of the contemplated power reactor. The design concepts are reviewed below:

1) The main component of a fast neutron reactor is a large stainless steel tank (like a deep swimming pool) that contains the primary liquid sodium (Na). For fire safety the top surface of the liquid sodium is covered by steel floats and over the floats there is an argon cover gas.

2) For safety there are no penetrations through the bottom or the vertical side walls of the liquid sodium tank;

3) The liquid sodium thermally stratifies so that the hotest liquid sodium floats on the top of the primary liquid sodium pool;

6) The active fuel bundle control portions consist of 244 fuel tubes which are raised and lowered by a liquid sodium hydraulic piston. The piston rings and piston tube providing the hydraulic seal are located outside the fast neutron flux.

7) Each fuel bundle has an associated: 7.5 m long indicator tube. Each indicator tube is _______ inch OD X ________inch wall steel tube with closed ends. At the inside bottom of the indicator tube is a pond of liquid mercury the vapor pressure of which indicates the fuel bundle discharge temperature. This temperature is used for fine adjustment of the active fuel bundle control portion position setpoint. This vapor pressure indicates temperature via a bourdon tube. Deflection of the bourdon tube causes laser spot deflection. The height of the indicator tube top indicates the control portion vertical position. The indicator tube vertical position and the active fuel bundle liquid sodium discharge temperature are acquired via laser optics.

14) Each fuel bundle bottom grating is formed from 48 4 inch X 1/8 inch steel strips that have saw cut notches so that the steel forms a square grating similar to the dividers separating wine bottles in a box of wine. There are welds at each metal junction for strength and rigidity. The fuel tube bottom plugs mate to the grating at each grating strip intersection. The gratings are welded to the fuel bundle girders. The bottom grating must reliably support the entire weight of the fuel tubes above the grating. The openings in this grating are 0.5 inch X 0.5 inch and allow liquid sodium to flow vertically between the fuel tubes. There is space under the bottom grating to allow liquid sodium cross flow in the event that the bottom grating is partially obstructed. The space between the fuel tubes is sufficient to allow liquid sodium cross flow if an individual cooling channel is obstructed.

16) Within the primary liquid sodium pool the active fuel bundles and the passive blanket fuel bundles are positioned in concentric octagons;

17) Outside the reactor core zone is a 1.60 m thick layer of neutron absorbing blanket fuel rods.
The design concept is to absorb all neutrons that escape from the blanket in the surrounding 3 m guard band of liquid sodium so that the neutrons do not activate or cause long term damage to the walls or bottom of the liquid sodium tank, the intermediate heat exchange bundles or the overhead gantry crane or the roof structure.

19) There is space in the primary liquid sodiuum pool for storage of neutron activated fuel bundles to allow fission products to naturally decay before these fuel bundles are removed from the liquid sodium pool.

20) The steel fuel bundle girders and shroud in and near the core zones are subject to intense fast neutron bombardment. These components are replaced with each fuel cycle. Hence, these components are designed for easy removal and reprocessing while still being highly radioactive;

21) The FNR is intended for partial refuelling every three years. The fuel bundles are designed to remain immersed in liquid sodium while they are transferred from their operating positions in the reactor to their cooling storage positions near the perimeter of the primary liquid sodium pool. These storage positions are outside the main neutron flux;

22) The active fuel bundle control portions can be withdrawn as a group by releasing the high pressure liquid sodium contained in the hydraulic actuators into the primary sodium pool. If a fuel bundle is running hotter than average its control portion position setpoint can be lowered.

23) On a loss of control system power all of the high pressure sodium is automatically released into the primary sodium pool to withdraw the active fuel bundle control portions to their cold shutdown position. The active fuel bundle control portion linear travel is about 1.2 m.

24) The actual vertical position of each active fuel bundle control portion is indicated by the height of the top of its indicator tube. The relative vertical position setpoint of each active fuel bundle control portion should be slowly adjusted to achieve the desired fuel bundle discharge temperature.

25) Each active fuel bundle acts as its own temperature control system. As the liquid sodium in the core zone warms up and thermally expands the fraction of fission neutrons diffusing out of the core zone increases, which reduces that zone's reactivity and thermal power output. Similarly as the liquid sodium contained in a core zone cools and contracts the fraction of fission neutrons diffusing out of that core zone decreases which increases that zone's reactivity and thermal power output. Hence, at a particular active fuel bundle control portion vertical position when the primary liquid sodium discharge temperature is high the fuel bundle thermal power output is low and as its primary liquid sodium discharge temperature decreases the fuel bundle thermal power output increases;

26) Thus every active fuel bundle in the fast neutron reactor spontaneously seeks an operating temperature at which the rate of heat generation equals the rate of heat removal by the natural convection flow of primary liquid sodium through the bundle. This rate is not uniform in the reactor because some fuel bundles will have been in the reactor longer than other fuel bundles, so the fuel tube swelling, fission product accumulation and dirt accumulation vary from bundle to bundle;

27) As long as the fuel in each bundle is uniform and its control portion is vertically positioned so that the safe liquid sodium discharge temperature of 445 degrees C is not exceeded the fast neutron reactor is passively thermally stable;

29) A square fuel tube lattice is used so that the natural convection liquid sodium flow is never dangerously reduced by fuel tube swelling. A square fuel tube lattice allows full rated power reactor operation with up to 15% linear fuel tube swelling.

33) In normal operation overall reactor thermal power is nearly constant or tracks the grid load. Thermal power turn down is achieved by allowing the primary liquid sodium temperature to rise causing chain reaction shutdowns in the active fuel bundle core zones. This temperature increase will occur on the reduction of the rate of heat transfer out of the reactor due to reduction of the secondary sodium flow rate;

34) An advantage of fast neutron reactors is that they are almost unaffected by slow neutron poisons. Hence the thermal power of a fast neutron reactor can ramp relatively rapidly to follow a changing electricity load.

35) Cold shutdown of a fast neutron reactor is achieved by withdrawing all the active fuel bundle control portions;

36) At full load the maximum permitted liquid sodium discharge temperature from any active fuel bundle is 440 degrees C. This temperature is chosen to keep the maximum fuel tube material temperature under 460 degrees C.

37) For safety the liquid sodium pool depth is made 16.5 m deep whereas the liquid sodium is only 15.5 m deep. Thus there is 1.0 m of tank depth allowance to withstand unforseen surface waves in the liquid sodium that might arise as a result of an earthquake. This allowance can also safely absorb Na from a intermediate heat exchange bundle leak.

38) The 3 m high straight inside sheet metal walls should be sufficiently reinforced to contain the large liquid sodium wave that might be produced by a large earthquake.

39) The insulated inner walls and inner ceiling of the reactor building confine the primary sodium in the event of a really large earthquake.

40) It is important to constantly filter the primary liquid sodium to keep the liquid sodium clean to prevent buildup of impurities on heat exchange surfaces or obstruction of the liquid sodium flow channels between the fuel tubes and between the intermediate heat exchange tubes.

41) An important issue is making the fuel tube gas plenum sufficiently large to safely contain both the spare sodium and the inert gas fission products.

42) The service life of the intermediate heat exchange bundles is long because the liquid sodium guard band protects them from cumulative neutron damage and primary sodium filtering minimizes surface deposits.

43) The service life of the primary liquid sodium pool liner is very long because there are no relevant corrosion or erosion mechanisms.

NEUTRON STOPPING CONSTRAINTS:
Two very important material constraints are that the probability of fast neutrons being absorbed by the 1.2 m thick blanket is close to unity and the probability of absorption of neutrons that escape from the blanket into the surrounding liquid sodium guard band is close to unity.

FUEL BUNDLE TRANSPORTATION:
A practical constraint on FNR design is transportation of fuel bundles back and forth between the FNR and the nearby fuel bundle assembly-disassembly facility. This transportation must be by truck. There are both fuel bundle length and weight constraints imposed by this transportation mode and the related shielding requirement.

Another consideration is that each fuel bundle is fabricated from HT-9 steel tubes, each 6.1 m long, so availability of suitable steel tubing at a competitive price is an important consideration.

The weight of the contained fuel bundle is additional. Thus the load is about 65 tonnes.

This weight is close to the maximum that can be easily transported by road. Hence there is no merit in contemplating a larger fuel bundle. Note that a fuel bundle has a chimney and an indicator tube that are added on at the reactor site.

ISSUES AFFECTING REACTOR SIZE:
1) From the perspective of reliable heat transport and reliable power generation it is essential for a FNR to have multiple independent intermediate heat transport and electricity generation systems so that a shutdown of one such system has only a small affect on the remaining electric power generation capacity. The present intent is to have 32 intermediate heat transport systems so as to make each electricity generator have a rated electricity output capacity of about 10 MWe.

2) The reactor fuel tube height is 6.0 m.

3) There is a reactor core zone height related to average reactor core zone fuel density that is necessary to realize core criticality. With 20% Pu fuel that core zone height works is about 0.35 m. Provision for fuel expansion due to fission product formation may eventually increase the reactor core zone height to about:
(0.35 m / 0.86) = 0.40 m.

The fuel bundle support frame which rests on the bottom of the primary liquid sodium pool is shipped to the site in multiple numerically machined parts.

In the passive fuel tubes the fuel rod stack is 3.731 m high. In the active tubes the fuel rod stack is initially 3.55 m high but in use can swell to 3.6 m.

There is 0.1 m of fuel tube length allocated to the two end plugs. There is 2.3 m of fuel tube length allowance for plenum. Hence the steel fuel tube length of 6 m is fully allocated.

To minimize the roof and gantry crane construction costs it is desirable to minimize the sodium pool width so as to minimize the required unsupported crane and roof spans.

REACTOR CORE:
In choosing the gap between the steel fuel tubes the issue is one of maximizing the average fuel density in the fuel bundle while not unduely decreasing the liquid sodium circulation and not imposing unreasonable cleanliness restrictions on the liquid sodium pool.

A related issue is that the maximum sodium temperature as liquid sodium passes through the fuel bundle needs to be limited to provide sufficient temperature safety margin at the upper end of the liquid sodium operating temperature range.

Based on all of these issues the center to center spacing between the square lattice fuel tubes was chosen to be:
(5 / 8) inch = 0.625 inch.

This dimensional choice sets the smallest initial intertube gap in the assembly at (1 / 8) inch, so filtering should be used to eliminate particulates larger than (1 / 32) inch in longest dimension.

Each fuel tube position has associated with it a reactor top surface area of about
1 tube per [0.625^2 inch^2]
= 1 tube / .3906 inch^2

PRIMARY SODIUM NATURAL CIRCULATION:
The natural circulation of the primary liquid sodium occurs due to a decrease in liquid sodium density with increasing temperature. Nuclear heating of the sodium in the reactor causes the sodium to locally expand in the core zones. If this expansion takes place within a surrounding pool of cooler liquid sodium the buoyancy of the warmer liquid sodium will cause it to rise. This warm liquid sodium flows over the top surface of the pool toward the ends of the pool where it cools, contracts and sinks as it flows between the cooler heat exchange tubes. The higher density cooled liquid sodium flows along the bottom of the primary sodium pool back to the pool bottom center where it again rises due to heating by the reactor fuel tubes.

In order to naturally circulate the primary liquid sodium there must be a large temperature difference between the top and bottom of the liquid sodium pool. At full power the bottom of the transition region between the hot liquid sodium and the cool liquid sodium should be at the top of the fuel tube chimneys. At zero thermal power the bottom of this transition region is 1.6 m above the bottom of the fuel tubes.

An accurate closed form expression for the maximum primary liquid sodium flow is developed at FNR PRIMARY LIQUID SODIUM FLOW. This viscous flow limits the FNR power output.

At steady state conditions the primary sodium mass flow through the reactor should equal to the primary sodium mass flow through the intermediate heat exchanger. Hence if the reactor thermal power is 1000 MWt and the temperature drop across the intermediate heat exchanger primary is 100 deg C, then the temperature rise along the reactor fuel tubes is:
100 deg C (1.00229) = 100.229 deg C

UPPER TEMPERATURE LIMIT:
When the reactor is operating at full rated power the liquid sodium temperature discharged from the top of the active fuel tube bundles must be less than 910 F or 488 C (Til & Yoon Figure 7-2, P. 149).

REACTOR CORE DESCRIPTION:
The steel fuel tubes are 6.0 m long. The steel fuel tubes are sealed closed at both the top and bottom ends.

Each fuel bundle is supported by a square steel grating 15.0 inches to a side which is firmly attached to the four corner girders. There are concentric square rings of 0.5 inch OD vertical steel fuel tubes on 0.625 inch square centers. The fuel tubes of each fuel bundle are position stablized by the bottom grating and (1 / 16) inch diameter bent criss cross rods.

Each active fuel tube bundle has a sliding central portion which controls the core zone reactivity of that fuel bundle. The position of this control portion is set by liquid sodium pressure applied to the actuator such that when sodium pressure is lost the control portion falls into its fully extracted position, which reduces reactivity of the fuel bundle. At the time of fuel insertion into the reactor the control portion should be fully extracted.

Each tube bundle is laterally stabilized by is shroud and by adjacent fuel bundles. The tube bundles are placed in position by the gantry crane that spans the width of the pool. The weight of the tube bundle assemblies is borne by the pool floor. The tube bundles are repositioned or replaced from time to time using the gantry crane and remote manipulation.

The first step in tube bundle replacement is reactor shutdown and disconnection of the chimney and indicator tube. Then the gantry crane lifts the selected tube bundle by the height of its frame insertion (~ 2.0 m) before moving the bundle horizontally to a perimeter storage position. During this process the spent fuel bundle remains covered by about 4 m of liquid sodium. The spent fuel bundle is moved to a perimeter storage position where it is mounted on another support pipe. The spent fuel bundle remains for three years in the liquid sodium until it loses most of its fission product decay heat. Then the spent fuel bundle is lifted out of the primary sodium pool for fuel reprocessing. Note that to access interior fuel bundles it is necessary to temporarily move other fuel bundles out of the way. There must be enough spare fuel bundle mounting positions around the reactor perimeter to permit accessing central fuel bundles.

ACTIVE FUEL BUNDLE CONTROL PORTIONS AND INDICATOR TUBES:
Each active fuel bundle has associated with it a reactivity control portion which lifts the bottom of the indicator tube. During normal reactor operation this control portion is positioned by liquid sodium pressure applied to its actuator. There are piston rings to achieve a good sliding seal between the piston OD and the ID of the actuator hydraulic cylinder. By appropriate active fuel bundle control portion positioning the reactor thermal load is evenly distributed.

Each active fuel bundle control portion has an attached indicator tube which indicates both the vertical position of the control portion and the fuel bundle liquid sodium discharge temperature. The indicator tube also provides a gamma ray/neutron propagation path for indicating fuel bundle power. At the bottom of the indicator tube, immediately above the control portion, is a pond of liquid mercury which has a known stable vapor pressure versus temperature characteristic. At the top of the indicator tube is a Bourdon tube or like device which bends in response to the mercury vapor pressure deflecting an incident laser beam coming from nearly overhead. This laser beam deflection indicates the fuel bundle's liquid sodium discharge temperature. The inside of the indicator tube should be lined with a thermally insulating material to minimize mercury vapor condensation on the indicator tube inside walls.

MAXIMUM TEMPERATURE:
Assume that when the reactor is operating at full rated power the liquid sodium discharge temperature from the core fuel bundles should remain less than 910 F or 488 C (Til & Yoon Figure 7-2, P. 149).

|THERMAL ANALYSIS:
The contemplated FNR is rated at 1000 MWt using HT-9 fuel tubes with a theoretical stress safety margin of over 2:1. At full rated power the maximum core rod temperature should normally be about 505 C. The fuel tubes heat an atmospheric pressure primary liquid sodium coolant that at full load is everywhere less than 440 degrees C.

These temperatures allow for a 15.0 deg C temperature drop across each HT-9 steel fuel tube wall and a 45 deg. C temperature drop between the inner HT-9 steel wall and the interior of the core fuel rod. Note that within the reactor flow channels the liquid sodium flow is laminar, so there may be a 5 degree C temperature difference between the fuel tube outside wall temperature and the average passing liquid sodium temperature.

The primary liquid sodium pool heats an intermediate liquid sodium heat transport loop that normally operates from 320 C to 440 C.
In the steam generator of the contemplated FNR at full load the water temperature is about 320 C and the corresponding saturated steam pressure is about 11.25 MPa. This pressure is held constant by a discharge pressure regulating valve. If this valve fails to open the steam generator pressure relief valve should trip at ~ 12 MPA.

With practical turbogenerator equipment the average electrical generation efficiency is about half the Carnot efficiency or about 0.23. Under best case operation it is about 0.33.

Thus the average electricity output will be 230 MWe. The best case electrical output will likely be about 330 MWe. This electricity output will drop in the winter but rise in the summer as the thermal load temperature changes.

The pressure within each intermediate heat transport loop is controlled by the expansion tank argon head pressure that operates in the range 0.1 MPa to 12 MPa. The intermediate liquid sodium pressure is kept slightly higher than the steam pressure so that in the event of a steam generator tube rupture the potential sodium leakage flow from the intermediate loop to the water is minimized and hydrogen generation is confined to the water side of the steam generator which is easily vented.

PRIMARY LIQUID SODIUM POOL:
The peak temperature in the primary liquid sodium pool is about 488 degrees C. Due to this high temperature conventional cement materials that set up via absorption of water of hydration are unsuitable for thermally unprotected liquid sodium containment.

Ideally the cavity for the liquid sodium pool should be cut from bed rock. The entire cavity should be above the highest local water table. For certain long term safety the bed rock should have a melting point over 600 degrees C. Sedimentary rock may be unsuitable if it breaks down at FNR liquid sodium operating temperatures.

Near the cavity walls the drill holes and explosive used should be chosen to minimize cracking of the remaining rock. The bedrock cavity should be about 31 m diameter X 22 m deep to allow for a 10.5 m sodium pool inner radius, 3 m thickness of lava rock (basalt) insulation thickness, a 1 m air gap and a 1 m thick concrete wall. Any cracks discovered in the cavity must be water sealed with clay. The outside of the concrete below grade should be sealed with hot bitumen.

The water tightness of the bedrock cavity should be checked by temporarily filling the bedrock cavity with water. If there is any sign of water leakage the entire cavity wall should be be lined with clay.

The bedrock base is leveled with igneous rock gravel. On top of the gravel foundation is laid 1 m of concrete, then a layer of structural steel I beams 1 m thick that will support the liquid sodium pool, reactor and lava rock while permiting cooling air to circulate beside and beneath the pool to remove heat conducted through the lava rock.

The concrete side walls are then formed up to 5 m above the pool deck level.

On top of the supporting I beams is the stainless steel sheet that forms the outer liquid sodium containment floor. External vertical I beams with horizontal spreaders reinforce the pool outer vertical side walls. The pool outer sheet steel vertical side walls are precisely positioned using threaded rods with turnbuckles that are attached to the adjacent concrete face. The details of this steel wall construction are similar to the construction of the hull of a ship.

The water tightness and strength of the outer steel wall and its supporting members should be demonstrated by temporarily filling the outer stainless steel wall with water.

Inside the outer sheet steel liquid sodium containment wall is a 3 m thickness of saw cut interlocking low density lava rock blocks that provide the thermal insulation between the inner and outer steel walls of the liquid sodium pool. Suitable lava rock for forming these blocks is plentiful on the main island of Hawaii. The saw cutting of the lava rock must be numerically controlled to be dimensionally accurate so that there are no gaps between the lava rock blocks. If the inner stainless steel containment wall for liquid sodium leaks the amount of liquid sodium that flows into the space between the inner and outer steel walls must be low and the thermal leakage to the outer stainless steel liquid sodium containment wall must be minimal. The lava rock used should have a melting point over 600 degrees C and should not be reduced by hot liquid sodium.

Inside the 2 m of fire brick is the inner liquid sodium containment wall which is fabricated from stainless steel sheets.

The inner and outer sheet steel walls are tied to the fire brick via thin steel rods that penetrate several brick layers before reaching a tie plate situated between adjacent brick layers.

The water tightness and strength of the inner steel wall should be confirmed by temporarily filling the inner steel wall with water.

In normal operation the inner stainless steel sheet vertical walls are in tension. In the event of an inner wall failure the outer wall will become in tension. Hence both walls must be sufficiently thick to be rated for the hydrostatic stress potentially imposed by the liquid sodium.

In normal operation the outer steel wall and its components are only slightly above room temperature. We must be concerned about long term corrosion of the outer steel wall due to it being exposed to an ongoing flow of cooling air drawn from the outside.

Another potential concern is wall stress and wall movement during earth quakes. The steel rods between the outer wall reinforcing I beams and the concrete wall should be sufficiently strong and resiliant to be earthquake tolerant.

Thus the liquid sodium has four hydraulically tested containment barriers, the inner stainless steel wall, the outer stainless steel wall, the concrete wall and the bedrock or fill cavity.

REACTOR SECTION FLOOR LINER:
An FNR designed for utility power production has many thousands of fuel rods. Sooner or later through accident, negligence or malevolent behavior a defective fuel bundle and/or related reactivity control system will be loaded into the reactor. In these circumstances a significant concern is fuel melting. If a steel fuel tube fails high density fuel rods, droplets or pellets might sink through the liquid sodium and will collect on the pool floor. It is essential that this material does not accumulate together sufficiently to form a critical mass. Thus the tube bundle support frame should be covered with a liner that has a high neutron absorption cross section and has controlled size bumps or cavities so that a critical mass cannot form. There should be a practical means of selectively removing and cleaning portions of this liner.

PRIMARY SODIUM CONTAMINANT FILTERING:
A power FNR will contain about 5000 m^3 of primary liquid sodium. In spite of best efforts to prevent sodium contamination the hot liquid sodium will gradually become contaminated. Hence a dedicated sump pump is required to run continuously to pump liquid sodium out of the bottom of the primary sodium pool, through a cleanup filter apparatus and back into the pool.

Each reactor tube bundle contains numerous narrow internal sodium flow channels. Thus it is imperative that there be a filter process that constantly removes sodium contaminants. The primary sodium filter must trap and remove all particulate matter with any dimension in excess of (1 / 32) inch. This filter system must run for a long time before the steel tube bundles are inserted into the liquid sodium.

Over time the liquid sodium will gradually become polluted with unwanted material including magnesium particles, radio active fuel, other metals, sodium oxide, sodium hydroxide, etc. The contaminants will include:
Na2O, NaOH, Na3N, NaNO2, NaNO3

Na2O must be removed by density separation and filtering. It has a specific gravity of 2.27 and sublimates at 1275 degrees C.

NaOH melts at 318.4 degrees C and has a boiling point of 1390 degrees C. It has a specific gravity of 2.130. It can be removed by density separation and filtering at a temperature less than 318 degrees C. NaOH may tend to form on the surface of the cooler parts of the intermediate heat exchange bundles if the pressure in the steam generator falls below 11.25 MPa causing the temperature of the return secondary sodium to fall below 320 degrees C.

Na3N dissociates at 300 degrees C and hence nitrogen accumulates in the cover gas. At temperatures below 300 C the nitrogen will react with the sodium. Then Na3N can be remved by filtering.

NaNO2 MP = 271 degrees C, dissociates at 320 deg C, SG = 2.16, can be removed by filtering at less than 271 degrees C or by cryogenic separation of cover gas.

The primary liquid sodium temperature must drop below 300 degrees C to permit complete contaminent removal by filtering. Hence the filter system inlet should be at the intermediate heat exchanger primary sodium discharge where the primary liquid sodium temperature is lowest. At this point the pool floor should be deepest. To clean the intermediate heat exchange tubes it is necesary to run the reactor at part load to raise the intermediate sodium return temperature.

SAFETY SYSTEM CONCEPT:
For safety reasons the pressure of the argon gas over the secondary sodium is kept slightly higher than the steam pressure. Any unanticipated change in secondary sodium level in the secondary sodium expansion tank indicates a tube leak somewhere, as does any presence of hydrogen in either the steam / condensate circuit or the secondary sodium circuit. In response that heat transport circuit is shut down reactor is shut down and steam/hydrogen is released via the steam circuit vent. This safety concept keeps both water and hydrogen out of the secondary sodium circuit where they could potentially cause catastrophic damage to the pipes and/or intermediate heat exchanger by liquid sodium fluid hammer.

The reactor has 32 independent intermediate heat transport loops. In the event that one heat transport loop has any sort of fault that loop can be shut down and isolated while the other heat transport loops continue to operate. Each heat transport loop has its own intermediate heat exchange bundles, steam generator, and intermediate liquid sodium circulation pump. Each heat transport loop has its own turbogenerator, condenser, condensate injection pump. Cooling towers, transformers, switchgear and auxillary power are shared.

A steam pressure in excess of 12.0 MPa should also trigger a heat transport loop shut down.

The maximum permitted liquid sodium temperature in normal circumstances is 450 deg C. At that temperature the yield stress of the steel tubes and pipes starts to decrease. If the primary liquid sodium temperature exceeds 488 deg C at any point that temperature should trigger a cold reactor shutdown via control bundle withdrawal.

If for some reason the release of the control bundles fails to shut down the reactor and if the heat removal rate is insufficient the primary liquid sodium temperature will rise until thermal expansion of the core fuel bundles reduces the core reactivity, which will by itself shut down the reactor. However, even with a total fission shut down it is necessary to maintain sufficient cooling to remove fission product decay heat.

STRUCTURAL ISSUES:
As calculated on another web page the mass of each fuel bundle assembly is about 5 tonnes.

Clearly the gantry must be rated for at least 10 tonnes to allow for the circumstance when two or three fuel bundles stick together.

The load bearing capacity of shale bedrock is believed to be about 5000 kPa

Under the inner pool floor liner is a 2 m thickness of solid brick, then the outer pool liner, then a 1 m layer of I beams resting on a 1.0 m thick concrete/gravel foundation with a sump pit. The I beams must be long term protected from corrosion.

PLUTONIUM DOUBLING TIME:
An issue of great importance in large scale implementation of FNRs is the FNR run time required for one FNR to breed enough excess Pu-239 to allow startup of another identical FNR. This time may be calculated using the approximation that each plutonium atom fission releases of 3.1 neutrons of which 2.5 neutrons are required for sustaining reactor operation leaving 0.6 neutrons for breeding extra Pu-239. Thus one atom of Pu-239 has to fission to form 0.6 atoms of extra Pu-239.

This is the time required for one FNR to form enough excess Pu to allow starting another FNR. Clearly this doubling time is too long to enable rapid deployment of FNRs.

With large scale implementation of FNRs the available supply of plutonium and trans uranium actinides will soon be exhausted. Hence the issue of the Pu-239 doubling time physically constrains the rate of growth of the FNR fleet.

Thus FNRs are viable for disposing of transuranium actinides but due to the Pu-239 doubling time will not in the near future provide enough power capacity for complete displacement of fossil fuels.

FUEL BUNDLE TRACKING:
Each fuel bundle requires an individual code identifier on its indicator tube to allow tracking of its neutron exposure history.

FISSION PRODUCT DECAY HEAT REMOVAL:
One of the most important aspects of reactor design is provision for fission product decay heat removal under adverse circumstances. If an event occurs which causes a sudden reactor shutdown the reactor will continue to produce fission product decay heat at 5% to 10% of its full power rating. Hence it is essential to ensure ongoing removal of fission product decay heat under the most adverse circumstances.

Hence:
1) Under no circumstances, including a sodium pool inner wall leak, should the liquid sodium level ever fall to the point that the fuel tubes are not fully immersed in liquid sodium.
2) The gap and lava rock fill between the inner and outer pool walls must be designed such that if the inner wall fails and the liquid sodium leaks into the space between the two walls, the sodium pool surface level will not drop below the tops of the upper blanket rods.
3) There must be sufficient natural intermediate liquid sodium circulation to remove fission product decay heat from the primary liquid sodium pool;
4) In the event of an intermediate liquid sodium circuit fault it is essential that reactor cooling be maintained. Hence multiple redundant intermediate heat transport systems are required. The current design contemplates 32 independent heat removal systems.

PRIMARY SODIUM TEMPERATURE MAINTENANCE:
To prevent prolonged equipment restart problems due to sodium freezing it is essential to keep the primary liquid sodium above 100 degrees C at all times. Hence, there should be at least four 0.5 MWt oil fired boilers on site connected to separate intermediate liquid sodium circuits to ensure maintenance of the primary liquid sodium pool temperature.

Thus when operating at full rated power it will take about one half hour to raise the reactor from an off condition at 120 degrees C to an on condition at an average temperature of 400 degrees C.

Some important physical properties of water, sodium and argon, are:

PROPERTY

WATER

SODIUM

ARGON

Liquid Thermal Conductivity:

0.58 W / m-deg C

142 W / m-deg C

Density Rho:

1.0 kg / lit

.927 kg / lit

1.784 g / lit@101.025 KPa, 0 deg C

(1 / Rho) dRho / dT:

2.71 X 10^-4 / deg K

Heat Capacity (J / mol deg K):

75.24

28.230

20.786

Heat of Vaporization@101 kPa:

40.68 kJ / mole

97.42 kJ / mole

Molecular Weight (gm / mole):

18

22.9897

39.948

Viscosity Muv (kg / m-s):

7 X 10^-4

Melting Point@101kPa (deg C):

0

97.72

Boiling Point@101 kPa (deg C):

100

883

Vapor Pressure@46 C:

10.094 kPa

Vapor Pressure@70 C:

31.176 kPa

Vapor Pressure@100 C:

101.32 kPa

Vapor Pressure@134 C:

303.93 kPa

Vapor Pressure@180 C:

1001.9 kPa

Vapor Pressure@234 C:

3005.9 kPa

Vapor Pressure@281 C:

6510.5 kPa

1 Pa

Vapor Pressure@311 C:

9995.8 kPa

Vapor Pressure@344 C:

15.342 MPa

10 Pa

STEAM GENERATOR CONNECTION:
A 16.0 inch OD (12.812 inch ID) liquid sodium pipe from the intermediate heat exchanger passes through the reactor building side wall, runs straight and then does a 90 degree curve to feed its respective steam generator. Each such 16 inch pipe feeds approximately 1452 X 0.500 inch OD X .065 inch wall tubes in each steam generator. The two 36 inch diameter steam generator sections are assembled vertically one above the other so that steam easily rises to the top. Thus at least 8 m _____outside the concrete wall of the reactor building is fully occupied by steam generators, recirculation pumps and related equipment.

An aisle must be left clear to allow steam generator access, removal and replacement. Thus around the reactor building is an upper level perimeter space dedicated to steam generators, expansion tanks and recirculation pumps. Below that space is room for turbogenerators and condensers. Including flanges each steam generator requires about a 2.0 m _____length allocation.

Note that The steam generators likely require internal floating tube manifolds to minimize longitudinal thermal stress. If both the tubes and the shell are fabricated from Inconel 600 the steam generators are expensive.

At the water inlet and steam output ports on the shell side the 3 foot diameter steam generator must be heavily reinforced. The shell must safely contain the steam pressure. The steam pressure compresses the 0.500 inch OD tubes which operate at a slightly higher internal pressure.

The connections to the tube side of the steam generators must be high pressure liquid sodium tight. The whole issue of liquid sodium tight gaskets that continuously operate at high temperatures and pressures needs further investigation.

INTERMEDIATE SODIUM VOLUME:
The volume of each intermediate sodium circuit can be estimated by assuming that everywhere along that circuit the cross sectional area is approximately the same as the cross sectional area of a 12.8 inch inside diameter pipe.